1. Field of the Invention
This invention relates to high nickel austenitic stainless steels having an excellent resistance to degradation by neutron irradiation, which are used as structural materials for nuclear power plants of light-water reactors.
2. Description of the Related Art
Up to the present time, it has been known that when austenitic stainless steels such as SUS 304, 316, etc., having been used as structural materials for nuclear power plants of light-water reactors, are used for a long time and subjected to neutron irradiation of at least 1.times.10.sup.21 n/cm.sup.2 (E&gt;1 MeV), Cr depletes and Ni, Si, P, S, etc. enrich, at crystal grain boundaries, resuiting in tendency of causing stress corrosion cracking (SCC) in the presence of a high load stress in an environment of light-water reactors. This is called "irradiation assisted stress corrosion cracking" (IASCC). It has eagerly been desired to develop materials with low IASCC susceptibility, but such low IASCC susceptibility materials (excellent resistance to degradation by neutron irradiation) have not been developed yet.
Austenitic stainless steels such as SUS 304, 316, etc., have been used as structural materials for nuclear power plants of light-water reactors, but when these materials are subjected to neutron irradiation of at least 1.times.10.sup.21 n/cm.sup.2 (E&gt;1 MeV) by using for a long time, change of concentrations of elements further proceeds, which do not or hardly occurs before using. That is, it is known that when Cr depletes and Ni, Si, P, S, etc. enrich at crystal grain boundaries (which will hereinafter be referred to as "radiation induced segregation") and there is a high load stress or residual stress, the stress corrosion cracking (irradiation assisted stress corrosion cracking, IASCC) tends to occur in the high temperature and pressure water of a neutron irradiation environment in light-water. Furthermore, it is known that the presence of oxygen in a large amount in high temperature and pressure water accelerates generation of IASCC.
Thus, the inventors have made various studies on properties of stainless steels and as a result of comparison of the inventors' calculation results on the change quantity of concentrations of Cr and Ni at crystal grain boundaries, based on S. Dumbill and W. Hanks' measured values of the crystal grain boundary segregation of neutron irradiated materials (Sixth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 1993, p. 521) with the inventors' accumulated SCC test results of neutron-irradiated SUS 304, 316, etc., it is found that the above described IASCC occurs when, at grain boundaries after neutron irradiation, the amount of Cr is at most 15% and the amount of Ni is at least 20%, as shown in FIG. 2, in which slant line part shows an occurrence zone of SCC.
The inventors have considered that such a phenomenon of occurrence of IASCC is due to the fact that concentrations of elements at crystal grain boundaries are similar to a composition of Alloy 600 (NCF 600 of JIS). Namely, IASCC is considered to be probably due to the fact that compositions at crystal grain boundaries get low Cr and high Ni by the neutron irradiation thereby approaching the composition of Alloy 600 (non-irradiated material) and resulting in stress corrosion cracking (PWSCC: primary water stress corrosion cracking) in water at a high temperature and pressure which often takes place in Alloy 600. At the present time, however, the mechanism of occurrence of PWSCC in Alloy 600 has not been elucidated.
The inventors have made studies based on the above described knowledge and reached the present invention by specifying a composition of a suitable material and simultaneously, combining it with a heat treatment and post working method for rendering optimum a crystal form in an alloy.
That is to say, present invention aims at providing structural materials having a resistance to degradation by the neutron irradiation, resulting in no SCC in environments of light-water reactors (in high temperature and pressure water or in high temperature and pressure water saturated with oxygen) even after subjecting the materials to neutron irradiation of approximately at least 1.times.10.sup.22 n/cm.sup.2 (E&gt;1 MeV). This corresponds to the quantity of maximum neutron irradiation received up to the end of the plant life of light-water reactors and having a thermal expansion coefficient approximately similar to that of SUS 304, 316, etc.